Generation of a High Temperature Material Data Base and its Application to Creep Tests with French or German RPV-steel

Considering the hypothetical core melt down scenario for a light water reactor (LWR) a possible failure mode of the reactor pressure vessel (RPV) and its failure time has to be investigated for a determination of the loadings on the containment. Numerous experiments have been performed accompanied with material properties evaluation, theoretical, and numerical work /REM 1993/, /THF 1997/, /CHU 1999/. For pre- and post-test calculations of Lower Head Failure experiments like OLHF or FOREVER it is necessary to model creep and plasticity processes. Therefore a Fi-nite Element Model is developed at the FZR using a numerical approach which avoids the use of a single creep law employing constants derived from the data for a limited stress and temperature range. Instead of this a numerical creep data base (CDB) is developed where the creep strain rate is evaluated in dependence on the current total strain, temperature and equivalent stress. A main task for this approach is the generation and validation of the CDB. Additionally the implementation of all relevant temperature dependent material properties has been performed. For an evaluation of the failure times a damage model according to an approach of Lemaitre is applied. The validation of the numerical model is performed by the simulation of and com-parison with experiments. This is done in 3 levels: starting with the simulation of sin-gle uniaxial creep tests, which is considered as a 1D-problem. In the next level so called "tube-failure-experiments" are modeled: the RUPTHER-14 and the "MPA-Meppen"-experiment. These experiments are considered as 2D-problems. Finally the numerical model is applied to scaled 3D-experiments, where the lower head of a PWR is represented in its hemispherical shape, like in the FOREVER-experiments. This report deals with the 1D- and 2D-simulations. An interesting question to be solved in this frame is the comparability of the French 16MND5 and the German 20MnMoNi55 RPV-steels, which are chemically nearly identical. Since these 2 steels show a similar behavior, it should be allowed on a lim-ited scale to transfer experimental and numerical data from one to the other

Thermo-mechanische Finite-Elemente-Modellierung zur Schmelzerückhaltung im RDB nach Verlagerung von Corium in das untere Plenum Thermo-mechanical finite element modelling of in-vessel melt retention after corium relocation into the lower plenum

Bezüglich eines hypothetischen Kernschmelzeszenarios in einem Leichtwasserreaktor ist es notwendig, mögliche Versagensformen des Reaktordruckbehälters sowie Versagenszeiträume zu untersuchen, um die Belastung für das Containment bestimmen zu können. Vom Institut für Sicherheitsforschung des FZD wurden Finite-Elemente-Modelle erstellt, die sowohl die Temperaturfeldberechnung für die Wand als auch die elastoplastische Mechanik der Behälterwand beschreibt. Die thermischen und mechanischen Berechnungen sind gekoppelt. Das Modell ist in der Lage, Versagenszeit und Versagensposition eines Behälters mit beheiztem Schmelzepool zu berechnen. Es existieren Modelle für die Druckwasserreaktortypen KONVOI und WWER-1000. Es wurden prototypische Szenarien mit und ohne externe Flutung des RDB untersucht, wobei die homogen und die segregierte Schmelzepoolkonfiguration betrachtet wurden. Zusätzlich wurde eine bruchmechanische Bewertung des Thermoschocks, der durch die externe Flutung entsteht, vorgenommen. Auf Grundlage der Experimente im Rahmen des ISTC-Projekts METCOR wurde außerdem die Auswirkung der thermochemischen Wechselwirkung zwischen Corium-Schmelze und RDB-Wand auf das Versagensverhalten des RDB untersucht. Das wichtigste Ergebnis ist, dass eine erfolgreiche Schmelzerückhaltung im RDB auch bei größeren Reaktoren möglich erscheint, wenn eine rechtzeitige Flutung der Reaktorgrube gelingt. Mittels einer statistischen Analyse wurden die Empfindlichkeiten von Ergebnissen gegenüber den Eingangsparametern und die Unsicherheiten der Ergebnisse quantifiziert. Considering the hypothetical core melt down scenario for a light water reactor (LWR) a possible failure mode of the reactor pressure vessel (RPV) and its failure time has to be investigated for a determination of the loadings on the containment. Several experiments have been performed accompanied with material properties evaluation, theoretical, and numerical work. At the Institute of Safety Research of the FZD finite element models have been developed simulating the thermal processes and the viscoplastic behaviour of the vessel wall. The thermal hydraulic and the mechanical calculations are coupled. The model is capable of evaluating fracture time and fracture position of a vessel with an internally heated melt pool. Models exist for the pressurised water reactor types KONVOI and VVER-1000. Prototypic scenarios with and without external flooding were investigated with consideration of homogeneous and segregated melt pool configurations. Additionally a fracture mechanic evaluation of the thermal shock, originating from the external flooding, was performed. Based on the experimental results of the ISTC project METCOR, the effects of the thermal chemical interaction between corium melt and vessel steel were investigated in the IVR scenarios. An important result of the project is that a successful in-vessel melt retention seems to be possible even for large reactors if the reactor pit can be filled with water before the corium melt is relocated to the lower plenum. By means of statistical analysis the sensitivity of results against input parameter variations was studied. The uncertainty of results was quantified

Thermomechanische Modellierung eines Reaktordruckbehälters in der Spätphase eines Kernschmelzunfalls

Considering the late in-vessel phase of an unlikely core melt down scenario in a light water reactor (LWR) with the formation of a corium pool in the lower head of the re-actor pressure vessel (RPV) the possible failure modes of the RPV and the time to failure have to be investigated to assess the possible loadings on the containment. In this work, an integral model was developed to describe the processes in the lower plenum of the RPV. Two principal model domains have to be distinguished: The temperature field within the melt and RPV is calculated with a thermodynamic model, while a mechanical model is used for the structural analysis of the vessel wall. In the introducing chapters a description is given of the considered accident scenario and the relevant analytical, experimental, and numerical investigations are discussed which were performed worldwide during the last three decades. Following, the occur-ring physical phenomena are analysed and the scaling differences are evaluated between the FOREVER-experiments and a prototypical scenario. The thermodynamic and the mechanical model can be coupled recursively to take into account the mutual influence. This approach not only allows to consider the tem-perature dependence of the material parameters and the thermally induced stress in the mechanical model, it also takes into account the response of the temperature field itself upon the changing vessel geometry. New approaches are applied in this work for the simulation of creep and damage. Using a creep data base, the application of single creep laws could be avoided which is especially advantageous if large temperature, stress, and strain ranges have to be covered. Based on experimental investigations, the creep data base has been de-veloped for an RPV-steel and has been validated against creep tests with different scalings and geometries. It can be stated, that the coupled model is able to exactly describe and predict the vessel deformation in the scaled integral FOREVER-tests. There are uncertainties concerning the time to failure which are related to inexactly known material parame-ters and boundary conditions. The main results of this work can be summarised as follows: Due to the thermody-namic behaviour of the large melt pool with internal heat sources, the upper third of the lower RPV head is exposed to the highest thermo-mechanical loads. This region is called hot focus. Contrary to that, the pole part of the lower head has a higher strength and therefore relocates almost vertically downwards under the combined thermal, weight and internal pressure load of the RPV. On the one hand, it will be possible by external flooding to retain the corium within the RPV even at increased pressures and even in reactors with high power (as e.g. KONVOI). On the other hand, there is no chance for melt retention in the considered scenario if neither internal nor external flooding of the RPV can be achieved. Two patents have been derived from the gained insights. Both are related to pas-sively working devices for accident mitigation: The first one is a support of the RPV lower head pole part. It reduces the maximum mechanical load in the highly stressed area of the hot focus. In this way, it can prevent failure or at least extend the time to failure of the vessel. The second device implements a passive accident mitigation measure by making use of the downward movement of the lower head. Through this, a valve or a flap can be opened to flood the reactor pit with water from a storage res-ervoir located at a higher position in the reactor building. With regard to future plant designs it can be stated - differing from former presump-tions - that an In-Vessel-Retention (IVR) of a molten core is possible within the reac-tor pressure vessel even for reactors with higher power

Thermomechanische Modellierung eines Reaktordruckbehälters in der Spätphase eines Kernschmelzunfalls

Considering the late in-vessel phase of an unlikely core melt down scenario in a light water reactor (LWR) with the formation of a corium pool in the lower head of the re-actor pressure vessel (RPV) the possible failure modes of the RPV and the time to failure have to be investigated to assess the possible loadings on the containment. In this work, an integral model was developed to describe the processes in the lower plenum of the RPV. Two principal model domains have to be distinguished: The temperature field within the melt and RPV is calculated with a thermodynamic model, while a mechanical model is used for the structural analysis of the vessel wall. In the introducing chapters a description is given of the considered accident scenario and the relevant analytical, experimental, and numerical investigations are discussed which were performed worldwide during the last three decades. Following, the occur-ring physical phenomena are analysed and the scaling differences are evaluated between the FOREVER-experiments and a prototypical scenario. The thermodynamic and the mechanical model can be coupled recursively to take into account the mutual influence. This approach not only allows to consider the tem-perature dependence of the material parameters and the thermally induced stress in the mechanical model, it also takes into account the response of the temperature field itself upon the changing vessel geometry. New approaches are applied in this work for the simulation of creep and damage. Using a creep data base, the application of single creep laws could be avoided which is especially advantageous if large temperature, stress, and strain ranges have to be covered. Based on experimental investigations, the creep data base has been de-veloped for an RPV-steel and has been validated against creep tests with different scalings and geometries. It can be stated, that the coupled model is able to exactly describe and predict the vessel deformation in the scaled integral FOREVER-tests. There are uncertainties concerning the time to failure which are related to inexactly known material parame-ters and boundary conditions. The main results of this work can be summarised as follows: Due to the thermody-namic behaviour of the large melt pool with internal heat sources, the upper third of the lower RPV head is exposed to the highest thermo-mechanical loads. This region is called hot focus. Contrary to that, the pole part of the lower head has a higher strength and therefore relocates almost vertically downwards under the combined thermal, weight and internal pressure load of the RPV. On the one hand, it will be possible by external flooding to retain the corium within the RPV even at increased pressures and even in reactors with high power (as e.g. KONVOI). On the other hand, there is no chance for melt retention in the considered scenario if neither internal nor external flooding of the RPV can be achieved. Two patents have been derived from the gained insights. Both are related to pas-sively working devices for accident mitigation: The first one is a support of the RPV lower head pole part. It reduces the maximum mechanical load in the highly stressed area of the hot focus. In this way, it can prevent failure or at least extend the time to failure of the vessel. The second device implements a passive accident mitigation measure by making use of the downward movement of the lower head. Through this, a valve or a flap can be opened to flood the reactor pit with water from a storage res-ervoir located at a higher position in the reactor building. With regard to future plant designs it can be stated - differing from former presump-tions - that an In-Vessel-Retention (IVR) of a molten core is possible within the reac-tor pressure vessel even for reactors with higher power

Generation of a High Temperature Material Data Base and its Application to Creep Tests with French or German RPV-steel

Considering the hypothetical core melt down scenario for a light water reactor (LWR) a possible failure mode of the reactor pressure vessel (RPV) and its failure time has to be investigated for a determination of the loadings on the containment. Numerous experiments have been performed accompanied with material properties evaluation, theoretical, and numerical work /REM 1993/, /THF 1997/, /CHU 1999/. For pre- and post-test calculations of Lower Head Failure experiments like OLHF or FOREVER it is necessary to model creep and plasticity processes. Therefore a Fi-nite Element Model is developed at the FZR using a numerical approach which avoids the use of a single creep law employing constants derived from the data for a limited stress and temperature range. Instead of this a numerical creep data base (CDB) is developed where the creep strain rate is evaluated in dependence on the current total strain, temperature and equivalent stress. A main task for this approach is the generation and validation of the CDB. Additionally the implementation of all relevant temperature dependent material properties has been performed. For an evaluation of the failure times a damage model according to an approach of Lemaitre is applied. The validation of the numerical model is performed by the simulation of and com-parison with experiments. This is done in 3 levels: starting with the simulation of sin-gle uniaxial creep tests, which is considered as a 1D-problem. In the next level so called "tube-failure-experiments" are modeled: the RUPTHER-14 and the "MPA-Meppen"-experiment. These experiments are considered as 2D-problems. Finally the numerical model is applied to scaled 3D-experiments, where the lower head of a PWR is represented in its hemispherical shape, like in the FOREVER-experiments. This report deals with the 1D- and 2D-simulations. An interesting question to be solved in this frame is the comparability of the French 16MND5 and the German 20MnMoNi55 RPV-steels, which are chemically nearly identical. Since these 2 steels show a similar behavior, it should be allowed on a lim-ited scale to transfer experimental and numerical data from one to the other

Generation of a High Temperature Material Data Base and its Application to Creep Tests with French or German RPV-steel

Considering the hypothetical core melt down scenario for a light water reactor (LWR) a possible failure mode of the reactor pressure vessel (RPV) and its failure time has to be investigated for a determination of the loadings on the containment. Numerous experiments have been performed accompanied with material properties evaluation, theoretical, and numerical work /REM 1993/, /THF 1997/, /CHU 1999/. For pre- and post-test calculations of Lower Head Failure experiments like OLHF or FOREVER it is necessary to model creep and plasticity processes. Therefore a Fi-nite Element Model is developed at the FZR using a numerical approach which avoids the use of a single creep law employing constants derived from the data for a limited stress and temperature range. Instead of this a numerical creep data base (CDB) is developed where the creep strain rate is evaluated in dependence on the current total strain, temperature and equivalent stress. A main task for this approach is the generation and validation of the CDB. Additionally the implementation of all relevant temperature dependent material properties has been performed. For an evaluation of the failure times a damage model according to an approach of Lemaitre is applied. The validation of the numerical model is performed by the simulation of and com-parison with experiments. This is done in 3 levels: starting with the simulation of sin-gle uniaxial creep tests, which is considered as a 1D-problem. In the next level so called "tube-failure-experiments" are modeled: the RUPTHER-14 and the "MPA-Meppen"-experiment. These experiments are considered as 2D-problems. Finally the numerical model is applied to scaled 3D-experiments, where the lower head of a PWR is represented in its hemispherical shape, like in the FOREVER-experiments. This report deals with the 1D- and 2D-simulations. An interesting question to be solved in this frame is the comparability of the French 16MND5 and the German 20MnMoNi55 RPV-steels, which are chemically nearly identical. Since these 2 steels show a similar behavior, it should be allowed on a lim-ited scale to transfer experimental and numerical data from one to the other

Thermo-mechanische Finite-Elemente-Modellierung zur Schmelzerückhaltung im RDB nach Verlagerung von Corium in das untere Plenum Thermo-mechanical finite element modelling of in-vessel melt retention after corium relocation into the lower plenum

Bezüglich eines hypothetischen Kernschmelzeszenarios in einem Leichtwasserreaktor ist es notwendig, mögliche Versagensformen des Reaktordruckbehälters sowie Versagenszeiträume zu untersuchen, um die Belastung für das Containment bestimmen zu können. Vom Institut für Sicherheitsforschung des FZD wurden Finite-Elemente-Modelle erstellt, die sowohl die Temperaturfeldberechnung für die Wand als auch die elastoplastische Mechanik der Behälterwand beschreibt. Die thermischen und mechanischen Berechnungen sind gekoppelt. Das Modell ist in der Lage, Versagenszeit und Versagensposition eines Behälters mit beheiztem Schmelzepool zu berechnen. Es existieren Modelle für die Druckwasserreaktortypen KONVOI und WWER-1000. Es wurden prototypische Szenarien mit und ohne externe Flutung des RDB untersucht, wobei die homogen und die segregierte Schmelzepoolkonfiguration betrachtet wurden. Zusätzlich wurde eine bruchmechanische Bewertung des Thermoschocks, der durch die externe Flutung entsteht, vorgenommen. Auf Grundlage der Experimente im Rahmen des ISTC-Projekts METCOR wurde außerdem die Auswirkung der thermochemischen Wechselwirkung zwischen Corium-Schmelze und RDB-Wand auf das Versagensverhalten des RDB untersucht. Das wichtigste Ergebnis ist, dass eine erfolgreiche Schmelzerückhaltung im RDB auch bei größeren Reaktoren möglich erscheint, wenn eine rechtzeitige Flutung der Reaktorgrube gelingt. Mittels einer statistischen Analyse wurden die Empfindlichkeiten von Ergebnissen gegenüber den Eingangsparametern und die Unsicherheiten der Ergebnisse quantifiziert. Considering the hypothetical core melt down scenario for a light water reactor (LWR) a possible failure mode of the reactor pressure vessel (RPV) and its failure time has to be investigated for a determination of the loadings on the containment. Several experiments have been performed accompanied with material properties evaluation, theoretical, and numerical work. At the Institute of Safety Research of the FZD finite element models have been developed simulating the thermal processes and the viscoplastic behaviour of the vessel wall. The thermal hydraulic and the mechanical calculations are coupled. The model is capable of evaluating fracture time and fracture position of a vessel with an internally heated melt pool. Models exist for the pressurised water reactor types KONVOI and VVER-1000. Prototypic scenarios with and without external flooding were investigated with consideration of homogeneous and segregated melt pool configurations. Additionally a fracture mechanic evaluation of the thermal shock, originating from the external flooding, was performed. Based on the experimental results of the ISTC project METCOR, the effects of the thermal chemical interaction between corium melt and vessel steel were investigated in the IVR scenarios. An important result of the project is that a successful in-vessel melt retention seems to be possible even for large reactors if the reactor pit can be filled with water before the corium melt is relocated to the lower plenum. By means of statistical analysis the sensitivity of results against input parameter variations was studied. The uncertainty of results was quantified

Thermo-mechanische Finite-Elemente-Modellierung zur Schmelzerückhaltung im RDB nach Verlagerung von Corium in das untere Plenum Thermo-mechanical finite element modelling of in-vessel melt retention after corium relocation into the lower plenum

Bezüglich eines hypothetischen Kernschmelzeszenarios in einem Leichtwasserreaktor ist es notwendig, mögliche Versagensformen des Reaktordruckbehälters sowie Versagenszeiträume zu untersuchen, um die Belastung für das Containment bestimmen zu können. Vom Institut für Sicherheitsforschung des FZD wurden Finite-Elemente-Modelle erstellt, die sowohl die Temperaturfeldberechnung für die Wand als auch die elastoplastische Mechanik der Behälterwand beschreibt. Die thermischen und mechanischen Berechnungen sind gekoppelt. Das Modell ist in der Lage, Versagenszeit und Versagensposition eines Behälters mit beheiztem Schmelzepool zu berechnen. Es existieren Modelle für die Druckwasserreaktortypen KONVOI und WWER-1000. Es wurden prototypische Szenarien mit und ohne externe Flutung des RDB untersucht, wobei die homogen und die segregierte Schmelzepoolkonfiguration betrachtet wurden. Zusätzlich wurde eine bruchmechanische Bewertung des Thermoschocks, der durch die externe Flutung entsteht, vorgenommen. Auf Grundlage der Experimente im Rahmen des ISTC-Projekts METCOR wurde außerdem die Auswirkung der thermochemischen Wechselwirkung zwischen Corium-Schmelze und RDB-Wand auf das Versagensverhalten des RDB untersucht. Das wichtigste Ergebnis ist, dass eine erfolgreiche Schmelzerückhaltung im RDB auch bei größeren Reaktoren möglich erscheint, wenn eine rechtzeitige Flutung der Reaktorgrube gelingt. Mittels einer statistischen Analyse wurden die Empfindlichkeiten von Ergebnissen gegenüber den Eingangsparametern und die Unsicherheiten der Ergebnisse quantifiziert. Considering the hypothetical core melt down scenario for a light water reactor (LWR) a possible failure mode of the reactor pressure vessel (RPV) and its failure time has to be investigated for a determination of the loadings on the containment. Several experiments have been performed accompanied with material properties evaluation, theoretical, and numerical work. At the Institute of Safety Research of the FZD finite element models have been developed simulating the thermal processes and the viscoplastic behaviour of the vessel wall. The thermal hydraulic and the mechanical calculations are coupled. The model is capable of evaluating fracture time and fracture position of a vessel with an internally heated melt pool. Models exist for the pressurised water reactor types KONVOI and VVER-1000. Prototypic scenarios with and without external flooding were investigated with consideration of homogeneous and segregated melt pool configurations. Additionally a fracture mechanic evaluation of the thermal shock, originating from the external flooding, was performed. Based on the experimental results of the ISTC project METCOR, the effects of the thermal chemical interaction between corium melt and vessel steel were investigated in the IVR scenarios. An important result of the project is that a successful in-vessel melt retention seems to be possible even for large reactors if the reactor pit can be filled with water before the corium melt is relocated to the lower plenum. By means of statistical analysis the sensitivity of results against input parameter variations was studied. The uncertainty of results was quantified